During operation in a nuclear reactor in which the coolant and/or moderator is water, corrosion occurs on the outside of the cladding tubes of nuclear fuel rods assembled in nuclear fuel assemblies which limits the length of time that the fuel rods and fuel assembly are usable in the nuclear reactor. A nuclear fuel rod typically has a cladding tube which houses a stack of fuel pellets of sintered uranium oxide, plutonium oxide or a mixture of such oxide fuels, and end plugs which seal both the upper and lower ends of the cladding tube. The cladding tube functions, in part, to prevent contact and thereby prevent chemical reaction between the fuel pellets and the coolant moderator as well as to prevent contamination of the coolant moderator by radioactive fission products emitted from the fuel pellets. The loss of leak tightness of the cladding tube could contaminate the reactor and associated systems and interfere with plant operations.
The cladding tube is required to have excellent mechanical properties and high corrosion resistance in the environment and for the conditions expected during reactor operations. Common cladding materials include zirconium, zirconium alloys, and stainless steel. Zirconium alloys in which the major component is zirconium are widely used for the cladding tube. Two of the most commonly used zirconium alloys are Zircaloy 2 and Zircaloy 4 and are described in American Society for Testing and Materials standard B350-93 (1993), Standard Specification For zirconium and zirconium Alloy Ingots For Nuclear Application, compositions R60802 and R60804, respectively. Zircaloy 2 (composition R60802) is composed of from 1.20 to 1.70 weight percent tin, 0.07 to 0.20 weight percent iron, 0.05 to 0.15 weight percent chromium, 0.03 to 0.08 weight percent nickel, where the iron plus chromium plus nickel content is from 0.18 to 0.38 weight percent, and the balance is zirconium plus impurities. Zircaloy 4 (composition R60804) is composed of from 1.20 to 1.70 weight percent tin, 0.18 to 0.24 weight percent iron, 0.07 to 0.13 weight percent chromium, where the iron plus chromium content is 0.28 to 0.37 weight percent, and the balance is zirconium plus impurities. The maximum impurities for Zircaloy 2 and Zircaloy 4 are given in the following table which is from Table 1 of the ASTM B350-93 Standard.
TABLE I ______________________________________ MAXIMUM IMPURITIES, WEIGHT % R 60802 R 60804 ______________________________________ Aluminum 0.0075 0.0075 Boron 0.00005 0.00005 Cadmium 0.00005 0.00005 Carbon 0.027 0.027 Cobalt 0.0020 0.0020 Copper 0.0050 0.0050 Hafnium 0.010 0.010 Hydrogen 0.0025 0.0025 Oxygen * Magnesium 0.0020 0.0020 Manganese 0.0050 0.0050 Molybdenum 0.0050 0.0050 Nickel -- 0.0070 Niobium 0.010 0.010 Nitrogen 0.0065 0.0065 Silicon 0.012 0.0120 Tin -- -- Titanium 0.0050 0.0050 Tungsten 0.010 0.010 Uranium (Total) 0.00035 0.00035 ______________________________________ *When so specified in a purchase order, oxygen shall be determined and reported. Maximum or minimum permissible values, or both, shall be as specified.
Although several zirconium alloys such as Zircaloy 2 and Zircaloy 4 have excellent properties for use as a cladding material, they are subject to corrosion, and at high degrees of burn up or after long in-reactor residence time, the cladding tube of fuel rods for light water reactors made of such alloys may reach a state of accelerated corrosion. Accelerated corrosion leads to very rapid increases in oxide thickness, frequently well beyond the currently accepted limit of approximately 80 to 100 microns, and equally if not more importantly results in high hydrogen absorption by the cladding material which may lead to unacceptable loss of cladding ductility due to hydride formation.
During corrosion of the cladding, the reactor coolant water reacts at a slow rate with the zirconium metal to form an oxide layer upon the surface of the cladding. In this reaction, hydrogen is being formed and can enter the zirconium metal, as a result of a zirconium water reaction Zr+2H.sub.2 O.fwdarw.ZrO.sub.2 +4H, or more particularly, EQU Zr+2H.sub.2 O.fwdarw.ZrO.sub.2 +2[(p)2H(abs)+(1-p)H.sub.2 (gas)](1)
where H(abs) is the portion of the corrosion generated hydrogen that is absorbed by the zirconium metal, H.sub.2 (gas) is the portion of the corrosion generated hydrogen which is released into the reactor coolant water, and p is the pick up fraction or the fraction of hydrogen generated in the corrosion reaction that is absorbed by the zirconium metal. A portion of the hydrogen thus produced diffuses into the zirconium metal, and the rest is released into the reactor coolant water. At the temperature of the cladding during reactor operation, the hydrogen absorbed by the cladding initially is in solid solution in and diffuses within the zirconium metal, but after the hydrogen concentration reaches the solid solubility limit for hydrogen in the zirconium metal, the hydrogen precipitates within the zirconium metal as a distinct separate phase in the form of zirconium hydride, a compound of zirconium and hydrogen.
At high degrees of burn up or after long in-reactor residence time, and the accumulation of a large number of hydride precipitates in the zirconium metal, accelerated corrosion of the cladding occurs and the corrosion reaction changes from Equation 1 to: EQU Zr+4H.sub.2 O+ZrH.sub.2 .fwdarw.2ZrO.sub.2 +5[(p)2H(abs)+(1-p)H.sub.2 (gas)](2)
where the corrosion rate increases very rapidly and the reactor coolant water reacts with both the zirconium metal and with the zirconium hydride to form zirconium oxide and hydrogen.
The zirconium hydride precipitates are not however evenly distributed throughout the zirconium metal cladding since the solid solubility limit of hydrogen in zirconium metal is directly proportional to the cladding temperature which varies across the thickness of the cladding wall. In regions of the cladding having a lower temperature, less hydrogen stays in solid solution in the zirconium and more hydrogen precipitates to form hydrides whereas in regions of the cladding having higher temperature, more hydrogen stays in solid solution within the zirconium and less hydrogen precipitates to form zirconium hydride. Thus, on the outside surface of the cladding (which is in contact with the reactor coolant water) and in those inner portions of the cladding which are closer to the outside of the cladding and away from the nuclear fuel pellet, a greater amount of hydrogen precipitates to form zirconium hydride, whereas less zirconium hydride precipitates are formed near the inner portion of the cladding closer to the fuel pellet. Fuel cladding designs which utilize a single zirconium alloy in a single layer cladding tube are thus subject to a predisposition for hydride accumulation on the outer lower temperature portions of the fuel rod cladding. In fuel cladding designs which are a composite of two or more layers of zirconium and/or zirconium alloys which are typically bonded together to form the cladding tube, the outer layer of zirconium or zirconium alloy, which is at a lower temperature during reactor operations, is also subject to the above-described predisposition to hydride formation, whereas the inner zirconium or zirconium alloy layer, by virtue of being at a higher temperature during reactor operations, is predisposed to less hydride formation. Accordingly, at least for these reasons, the outer layer of composite cladding of nuclear fuel rods for pressurized water reactors has been designed and material has been selected to be highly corrosion resistant, whereas the inner zirconium alloy layer has been selected to provide desirable mechanical properties to the cladding such as high yield strength and low creep rate.
Similarly, for these same reasons, the outer layer of composite cladding of nuclear fuel rods for boiling water reactors has been designed and material has been selected to be highly corrosion resistant and provide desirable mechanical properties to the cladding whereas the inner zirconium or zirconium alloy layer has been selected to provide enhanced resistance to stress corrosion cracking which may result from pellet-cladding interaction and the release of fission products from the fuel pellets.
It would therefore be an advantage over prior art fuel rod designs to provide a nuclear fuel rod with a cladding which avoids or overcomes the above-mentioned disadvantages and which has improved corrosion resistance and improved resistance to hydride formation, particularly on the outer layer of the cladding tube, and improved resistance to accelerated corrosion, while maintaining the strength of the cladding tube.